1. Field of the Invention
The invention is in the field of fusion power generators, particularly those utilizing fusion reactors of the magnetic confinement type.
2. Description of the Prior Art
Prior art concepts with regard to utilization of fusion energy for the economic production of power have been premised upon an ultimate design of a large scale reactor able to produce the desired power and lasting a sufficiently long time to justify the large capital investment required to build the reactor. The economics of a large capital investment with a long reactor lifetime have been carried over from the fission reactor field as an inherent basis in the design of economic fusion power plants. Consequently, plasma temperatures and densities have been parameterized to yield a maximum wall loading of the first wall (vacuum wall surrounding the plasma) consistent with durability of wall materials and a long replacement time which is economically acceptable. Typically, a maximum wall loading of 1-3 MW/m.sup.2 has been thought reasonable with a minimum replacement time of approximately five years.
Consistent with the projected long life of the fusion reactor, the plasma core has traditionally been made large so as to allow large power output with low energy loadings on the first wall as well as for reasons of plasma confinement in the regimes of traditional interest. Furthermore, the plasma core has traditionally been surrounded directly with a thick material blanket region to absorb the plasma-generated neutron energy as well as to protect the large and expensive magnetic field windings surrounding the blanket. These large field windings, required to confine plasma in the plasma core, must be large enough to surround the plasma core. Traditionally, superconducting magnets have been utilized in order to reduce the power required to drive the magnetic coils, and the blanket thus served to remove the coils from the regime of high neutron fluxes and associated radiation damage to which the superconductors are susceptible. Such superconducting magnets have a limited magnetic field capability of between approximately 80 and 150 kilogauss. The maximum permissible density and temperature of the plasma is in turn dictated by the strength of the magnetic field possible which, because of the foregoing considerations has been limited to the maximum strength available from the superconducting magnets. Thus, traditional fusion device concepts have involved large plasma volumes, thick blankets of large volume, low first wall loadings, and the use of large, expensive superconducting magnets placed outside the regions of the blanket, plasma core, and any added auxiliary shielding.
In utilizing large volume experimental reactors of the tokamak-type, and in the conceptual design of practical large volume toroidal reactors, ohmic heating inherently plays a negligible role in the process of raising the plasma temperatures to values of thermonuclear interests. This is true because the current density which can be induced in any toroidal plasma configuration is proportional to the magnetic field divided by the major radius of the torus. For the fields attainable by superconducting magnets and the dimensions of traditionally envisioned toroidal devices, the current density is insufficient to yield significant ohmic heating of the plasma. Thus, in both the experimental and conceptual designs large sources of energetic beams of neutral particles have been utilized to provide power to the plasma on the order of tens to hundreds of megawatts. Neutral beam injection techniques require the utilization of large access ports to the plasma through the surrounding magnetic structure thus adding to the cost and complexity of any practical fusion power plant. Additionally, in order to ensure proper beam penetration to the center of the plasma column, operation of neutral beam injection devices has been limited to plasma densities the order of 10.sup.14 /cm.sup.3.
As experimental fusion devices, blankets have typically not been employed inasmuch as they are unnecessary to study many of the basic physical processes involved in the plasma such as plasma fusion ignition, confinement, plasma heating and fusion reaction studies. The tokamak has provided an experimental tool for testing the feasibility of plasma confinement and has been the subject of extensive experimentation, e.g., see "The Tokamak Approach in Fusion Research" by Bruno Coppi et al, Scientific American, July 1972, U.S. Pat. No. 3,778,343 and "Tokamak Experimental Power Reactor Conceptual Design", Vols. 1 and 2, ANL/CTR-76-3 (August 1976), all of which documents are incorporated herein by reference. One particular tokamak device, the Alcator, has been designed to achieve large plasma currents with high toroidal magnetic field strengths. Typically, plasma currents on the order of 100 kiloamps with field strengths up to 82 kilogauss have been obtained. In such experimental devices, plasmas with densities up to 9.times.10.sup.14 particles per cubic centimeter with temperatures up to 1 keV have been contained. However, the Alcator approach is not typical of the majority of prior art devices which have focused on toroidal devices of much lesser density, larger dimension, smaller magnetic fields and which require extensive auxiliary heating (generally by neutral beam injection) to strive for plasma ignition temperatures.
The approach of a very high yield, high density and a small compact device such as the Alcator has been considered in the prior art as limited to merely academic interest for purposes of physics studies of plasma behavior but has not been considered of interest for future applications to practical fusion power production.
Another experimental area that has been developed for the magnetic confinement of thermonuclear plasma is embodied in the stellarator concept. While in the tokamak, the confining magnetic field is partially produced by external coils and partially by the current induced in the plasma, in a stellarator, the confining field is produced only by external coils. Both the tokamak and the stellarator, however, may be considered forms of a toroidal plasma confinement device.